Microstructure evolution and degradation mechanisms of reactor internal steel irradiated with heavy ions

Borodin, O.V.; Брык, В.В.; Kalchenko, А.S.; Parkhomenko, A. A.; Shilyaev, B.A.; Tolstolutskaya, G.D.; Voyevodin, V.N.

doi:10.1016/j.jnucmat.2008.12.007

Cited by 12 publications

(5 citation statements)

References 5 publications

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“…Then 5 MeV Ni ++ irradiations were performed at Orsay JANNuS platform at 450 °C. The temperature shift with respect to PWR conditions is necessary for compensating flux effects [28][29][30][31][32]. The cumulative dose at the depth where samples are prepared is 5 dpa.…”

Section: Resultsmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

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Abstract:Even if temperature, pressure and chemistry of the cooling water are not very high and aggressive, materials used in PWRs (Pressurized Water Reactors) are exposed to different degradation mechanisms. One of the main goals of the research programs in this field is to develop physical model of the mechanisms down to the atomic scale. Such approach needs a clear description and understanding of the degradation mechanisms at the same small scale. This paper illustrates the benefits of different microscopies and of their last improvements up to the promising possibilities of monochromated and aberrations corrected TEM/STEM. A specific focus is placed on four different degradation mechanisms observed in austenitic stainless steel: irradiation ageing, corrosion fatigue, stress corrosion cracking and corrosion.

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Section: Resultsmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

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“…It is a van de Graf electrostatic accelerator of heavy ions with an external injector of 1.8 MeV Cr 3+ ions. This facility has been used to irradiate other alloys [12][13][14][15][16][17][18][19][20]. Even though ESUVI has the ability to simultaneously and accurately implant both helium and hydrogen in the region to be examined [18], these gases were not employed in the current study.…”

Section: Irradiation Proceduresmentioning

confidence: 99%

Ion-induced swelling of ODS ferritic alloy MA957 tubing to 500 dpa

Toloczko

Garner²,

Voyevodin

et al. 2014

Journal of Nuclear Materials

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100

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In order to study the potential swelling behavior of the ODS ferritic alloy MA957 at very high dpa levels, specimens were prepared from pressurized tubes that were unirradiated archives of tubes previously irradiated in FFTF to doses as high at 110 dpa. These unirradiated specimens were irradiated with 1.8 MeV Cr+ ions to doses ranging from 100 to 500 dpa and examined by transmission electron microscopy. No coinjection of helium or hydrogen was employed. It was shown that compared to several ferritic/martensitic steels irradiated in the same facility, these tubes were rather resistant to void swelling, reaching a maximum value of only 4.5% at 500 dpa and 450°C. In this fine-grained material, the distribution of swelling was strongly influenced by the presence of void denuded zones along the grain boundaries.

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“…We assume a temperature of 450 ˚C unless otherwise noted. These irradiation conditions are similar to those encountered in the experiments conducted at the Kharkov Institute of Physics and Technology [7,[12][13][14][15]21]. The damage and implanted Cr profiles were calculated using the Kinchin-Pease option in SRIM-2013 [24,25].…”

Section: Methodsmentioning

confidence: 99%

“…There has been a significant body of experimental work on RIS in ion irradiated Fe-Cr based alloys (e.g., [7,[12][13][14][15][18][19][20][21]) but it not clear from these works what happens to implanted ions in the entire damage region, particularly at very high doses where the implanted ions can significantly change composition. In general, it is not understood how the injected interstitials impact local alloy composition and RIS in the damage region.…”

Section: Introductionmentioning

confidence: 99%

Simulated spatial and temporal dependence of chromium concentration in pure Fe and Fe 14%Cr under high dpa ion irradiation

Vörtler

Mamivand

Barnard

et al. 2016

Journal of Nuclear Materials

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In this work we develop an ab initio informed rate theory model to track the spatial and temporal evolution of implanted ions (Cr + ) in Fe and Fe-14%Cr during high dose irradiation. We focus on the influence of the specimen surface, the depth dependence of ion-induced damage, the damage rate, and the consequences of ion implantation, all of which influence the depth dependence of alloy composition evolving with continued irradiation. We investigate chemical segregation effects in the material by considering the diffusion of the irradiation-induced defects. Moreover, we explore how temperature, grain size, grain boundary sink strength, and defect production bias modify the resulting distribution of alloy composition. Our results show that the implanted ion profile can be quite different than the predicted SRIM implantation profile due to radiation enhanced transport and segregation.

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Microstructure evolution and degradation mechanisms of reactor internal steel irradiated with heavy ions

Cited by 12 publications

References 5 publications

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Ion-induced swelling of ODS ferritic alloy MA957 tubing to 500 dpa

Simulated spatial and temporal dependence of chromium concentration in pure Fe and Fe 14%Cr under high dpa ion irradiation

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